ML19309H347
| ML19309H347 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/21/1980 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309H348 | List: |
| References | |
| NUDOCS 8005130085 | |
| Download: ML19309H347 (13) | |
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%,i, 800513008 6 UNITED STATES y-NUCLEAR REGULATORY COMMISSION j
j WASHING TON, D. C. 20555 s
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET N0. 50-321 EDWIN 1. HATCH NUCLEAR PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated July 9, 1979, as amended July 27, September 21, October 29, November 30, December 31, 1979 ami February 18, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facil ty will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering tue health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordan'ce with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-57 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 74, are hereby incorporated in the 1icense. The 1icensee shall operate the facility in accordance with the Technical Specifications.
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This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COP 94ISSION Ow==
W. P. Gam 11, Acting Assistant Director for Operating Reactor Projects Division of Operating Reactors
Attachment:
Changes to the Technical Specifications j
Date of Issuance:
April 21,1980 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 74 FACILITY OPERATING l.ICENSE NO. DPR-57 DOCKET N0. 50-321 The proposed changes to the Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:
t Remove Page Insert Page 3.10-5 3.10-5 3.10-7 3.10-7 3.10-9 3.10-9 3.10-10 3.10-10 5.0-1 5.0-1 5.0-2 5.0-2 i
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o LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.10.F.3 Monorail Hoist (Continued) a.
The trolley and hoist shall be demonstrated to be operable by a trial lift of the spent fuel pool gate or an equivalent weight.
b.
A visual inspection shall be made to insure the st actural integrity of the 5-to:1 mono-rail hoist.
3.10.G Spent Fuel Cask Lifting 4.10. G Spent Fuel Cask Lif ting Trunnions and Yoke Trunnions and Yoke See note for Specification See note for Specification 3.10.F.1 above.
4.10.F.1 above.
3.10.H Time Limitation Irradiated fuel shall not be handled in or above the reactor prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown.
3.10.1 Crane Travel-Spent Fuel 4.10.I Crane Travel-Spent Fuel Storage Pool Storage Pool Loads in excess of 1600 Loads, other than fuel assemblies pounds shall be prohibited
' or control rods, shall be verified from travel over fuel to be 51600 pounds prior to move-assemblies in the spent ment over fuel assemblies in the fuel storage racks.
fuel storage pool racks.
3.10-5 Amendment No. 74 T
3.10.A.2.
_Fual Grapple Ho' ' Load Setting Interlocks Fuel handling is normally conducted with the fuel grapple hoist. The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly. This total is approximately 1500 lbs.
in comparison to the load setting of 485 t 30 lbs.
3.
Auxiliary Hoists Load Setting Interlock Provisions have 'also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks. The 485130 lb.
load setting of these hoists is adequate to trip the interlock when a fuel bundle is being handled.
b.
Fuel Loading To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core.
This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent cri tical i ty.
C.
Core Monitoring During Core Alterations The SRM's are provided to monitor the core during periods of Unit shutdown and to guide the operator during refueling operations and Unit startup.
Requiring two operable SRM's in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant daring such al terations. The requirements of 3 counts per second provides assurance that neutron flux is being monitored.
During sprial unloading, it is not necessary to maintain 3 cps because core alterations will involve only reactivity removal and will not result in cri tical i ty.
The loading of diagonally adjacent bundles around the SRM's before attaining the 3 cps is permissible because these bundles were in a subcritical configuration when they were removed and therefore they will remain subcritical when placed back in their previous poritions.
D.
Spent Fuel Pool Water Leve' The design of the spent fuel storage pool provides a storage location for 3181 fuel assemblies in the reactor building which ensures adequate shielding, l
cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the active fuel will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10CFR20.
The normal water level provides 14-1/2 feet of additional water shielding.
All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the water level to less than 10 feet above the top of the active fuel.
Lines extending below this level are equipped with two check valves in series to prevent inadvertent pool drainage.
E.
Control Rod Drive Maintenance During certain periods, it is desirable to perform maintenance on two control rod drives at the same time.
Amendment No. 74 3.10-7
i BASES FOR LIMITING CONDITIONS FOR OPERATION t
3.10.H Time Limitation The radiological consequences of a fuel handling accident are based upon the ac'c16ent occurring at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown.
3.10.1 Crane Travel-Spc- _..el Storage Pool A maximum weight of 1600 pounds shall be permitted to be transported over stored spent fuel in order to minimize the consequences of a load handling accident.
3.10.J References 1.
FSAR Section 7.6, Refueling Interlocks 2.
FSAR Section 7.5, Neutron Monitoring System 3.
Morgan, W.
R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors," General Electric Company, Atomic Power Equipment Depa'rtment, November 1968, revised April 1969 (APED-5706) s 4.
FSI-R Section 10.3, Spent Fuel Storage 5.
FSAR Section 3.6.5.2, Reactivity Control F
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3.10-9
-Amendment No.
74
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O BASES FOR SURVEILLANCE REQUIREMENTS 4.10.
REFUELING A.
Refueling Interlocks Complete functional testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.
By loading each hoist with a weight equal to the fuel assembly, positioning the refueling platform, and withdrawing control rods, the interlocks can be subjected to valid operatiotial tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each re-dundant logic element can independently perform its functions.
C.
Core Monitoring During Core Alterations Requiring the SRM's to be functionally tested prior to any core alteration assures that the SRM's will be operable at the start of that alteration.
The daily response check of the SRM's ensures their continued operability.
D.
Spent Fuel Pool Water Level A daily record of the Spe'nt Fuel Pool Water Level to determine that the minimum of 8.5 feet is not exceeded is considered sufficient to ensure that radiological shielding is maintained.
E.
Control Rod Drive Maintenance Refueling interlocks and core monitoring surveillance are discussed in 4.10. A and 4.10.C above. The choice of the strongest (highest reactivity worth) rod which will be used for a determination of the relevant shutdown margins is based on prior core calculations supplemented by empirical data obtained from similar cores.
From similar data and calculations the reactivity worth of rods adjacent to a withdi'wn rod will also be known.
Thus the surveillance shutdown margins can be evaluated in terms of rod posi tion.
F.
Reactor Building Cranes Modifications to the main reactor building crane are being studied in order to increase its ability to withstand a single failure.'
A spent fuel cask will not be lif ted until these modifications have been accepted by the AEC and the AEC has approved the lif ting of casks by the crane and the appropriate Technical Specifications.
G.
Spent Fuel Cask Lif ting Trunnions and Yoke See note for Bases 4.10.F above.
I I.
Crane Travel-Spent Fuel Storage Pool Refer to Bases 3.10.I.
3.10-10 Amendment No. 74
5.0 MAJOR DESIGN FEATURES A.
Si te Edwin I. Hatch Nuclear Plant Unit No.1 is located on a site of about 2244 acres, which is owned by Georgia Power Company, on the south side of the Altamaha River in Appling County near Baxley, Georgia.
The Universal Transvedse Mercator Coordinates of the center of the reactor building are:
Zone 17R LF 372,935.2m E and 3,533,765.2m N.
B.
Reactor Core 1.
Fuel Assemblies The core shall consist of not more than 560 fuel assemblies of the licensed combination of 7x7 bundles which contain 49 fuel rods and 8x8 fuel bundles whkh contain 62 or 63 fuel rods each.
2.
Control Rods The reactor shall contain 137 cruciform-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 4
70% of its theoretical derisity.
C.
Reactor Vessel The reactor vessel is described in Table 4.2-2 of the FSAR.
The applicable design specifications shall be as listed in Table 4.2-1 of the FSAR.
D.
Containment 1.
Primary Containment The principal design parameters and characteristics of the primary con-tainment shall be as given in Table 5.2-1 of the FSAR.
2.
Secondary Containment The secondary conthinment shall be as described in Section 5.3.3.1 of the FSAR and the applicable codes shall be as given in Section 12.4.4 of the FSAR.
s 3.
Primary Containment Penetrations Penetrations to the primary containment and piping passing 'through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.
E.
Fuel Storage 1.
Spent Fuel All arrangements of fuel in the spent fuel storage racks shall be main -
n t greater than 0.95.
tained in a subcritical configuration having a keff 2.
New Fuel The new fuel storage vault shall be such that the k*ff dry shall not be flooded shall not be greater than 0.95.
greater than 0.90 ~and the kgff Amendment No. 74 5.0-1
3.
Fuel Storage Fuel in the Spent Fuel Pool shall have a maximum fuel loading of 15.2 grams of Uranium-235 per axial centimeter.
5.0.F.SeirbicDesign h
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The reactor building and al'1 engineered safeguard systems are designed for w
the design basis earthquake with a horizontal ground acceleration of 0.15 g.
The operating basis earthquake has a horizontal ground acceleration of l
0.08 g.
C. Re ferences
- 1. FSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design
- 2. FSAR Section 5.2, Primary Containment System
- 3. FSAR Section 5.3, Secondary Containment System
- 4. TSAR Section 12.4.4, Governing Codes and Regulations
- 5. FSAR Section 10.3, Spent Fuel Storage
- 6. FSAR Section 10.2, New Fuel Storage 4
j-5.0-2
' Asiendment No. 74
<?p nu c,, (o, UNITED STATES a
NUCLEAR REGULATORY COMMISSION h.
C WASHINGTON, D. C. 20555
,o GEORGIA POWER COMPANY OGLETHORPE ELECTRIC POWER CORPORATION MUNICIPAL ELECTRIC AUlHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.15 License No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated July 9, 1979, as amended July 27, 1979, September 21, October 29, November 30, December 31, 1979 and February 18, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;
B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commi ssion; i
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.15, are hereby incorporated in the license. 'The licensee shall operate the facility in accordance with the Technical Specifications.
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, 3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W ^ :--
W. P. Gammi31, Acting Assistant Director for Operating Reactor Projects Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April 21, 1900
i ATTACHMENT TO LICENSE AMENDMENT NO.15 FACILITY OPERATING LICENSE N0. NPF-b DOCKET NO. 50-366 4
The proposed changes to the Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:
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Insert Page 5-4 5-4 4
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DESIGN FEATURES ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 185 feet.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2845 fuel assemblies.
l FUEL STORAGE 5.6.4 Fuel in the Spent Fuel Pool shall have a maximum fuel loading of 15.2 grams of Uranium-235 per axial centimeter.
4 5.7 COMP 0NENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.
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Amendment No.
15.
HATCH -' UNIT 2 5-4 L
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